Ueda Japanese Patent Showa 52-50489 disclosed the use of part length fuel rods for the creation of an improved fuel to moderator ratio in the upper two phase region of a fuel bundle, especially in the cold shut down state of the reactor. A fuel bundle was disclosed in which clustered part length fuel rods defined a large, central, generally conical shaped void in the upper two phase region of the fuel bundle.
Two embodiments were disclosed in the Ueda reference. A first embodiment includes a large conical water rod occupying the large central conical void defined collectively by the part length rods. A second, and apparently preferred embodiment, disclosed the conical region otherwise unoccupied.
Regarding this latter design, testing has established that while nuclear improvements in the upper two phase region in the cold state might be realized, adverse heat transfer performance, especially in terms of adverse critical power may be realized by large central void regions in a boiling water reactor fuel bundle. Specifically, the large defined void results in vapor being concentrated to the region. Unfortunately, surrounding portions of the two phase region tends to flow into the steam vent area. This results in the diversion of significant amounts of liquid coolant away from the heated rod surfaces adjacent the void, this liquid coolant being entrained in the accelerated steam flow within the large defined conical void. There results a reduced flow adjacent the full length fuel rods which surround the large void. This reduced flow rate has a corresponding reduced critical power on the rod surfaces adjacent the void. Overall fuel bundle efficiency is reduced.
In Dix et al. U.S. Pat. No. 5,017,332 entitled Two-Phase Pressure Drop Reduction BWR Assembly Design, issued May 21, 1991 (formerly U.S. Pat. Application Ser. No. 07/176,975, filed Apr. 4, 1988), we maintained the nuclear benefits and removed the adverse thermal hydraulic effects by using generally smaller open flow channels dispersed across the two phase region of the fuel bundle lattice. The dispersed flow channels realized the natural tendency of the vapor phase of the two phase mixture to migrate ("drift") toward the low resistance flow paths where the realized flow was primarily vapor. It has been found that such dispersed flow paths are favorable for a BWR fuel assembly since preferential diversion of vapor away from the fueled rods has combined nuclear, stability, and thermal hydraulic advantages.